Nuclear reactors have their fuel contained in sealed cladding for the isolation of the nuclear fuel from the moderator/coolant system. The term cladding, as used herein, refers to a zirconium-based alloy tube composed of at least one metal in addition to the zirconium base. The cladding may be composed of more than one layer including a zirconium alloy substrate and an unalloyed zirconium barrier. Typically, the cladding is formed in the shape of a tube with the nuclear fuel contained in pellet form therein. These pellets are stacked in contact with one another for almost the entire length of each cladding tube, which cladding tube is in the order of 160 inches in length. Typically, the cladding tube is provided with springs for centering the fuel pellets and so-called "getters" for absorbing fission gases. Thereafter, the internal portions of the fuel rod are pressurized with various gases for optimum dissipation of gases produced from the fission reaction, and sealed at both ends.
Zirconium and its alloys, under normal circumstances, are excellent nuclear fuel cladding since they have low neutron absorption cross sections and at temperatures below about 398.degree. C. (at or below the core temperature of the operating reactor) are strong, ductile, stable, and nonreactive in the presence of demineralized water or steam. "Zircaloys" are a widely used family of corrosion-resistant zirconium alloy cladding materials. The Zircaloys are composed of 97-99% by weight zirconium, with the balance being tin, iron, chromium, and nickel. "Zircaloy-2" and "Zircaloy-4" are two widely-used zirconium-based alloys for cladding. Zircaloy-2 has on a weight basis about 1.2 to 1.7 percent tin; 0.13-0.20 percent iron; 0.06-0.15 percent chromium and 0.05 to 0.08 percent nickel. Zircaloy-4 has essentially no nickel and about 0.2% iron but is otherwise substantially similar to Zircaloy-2.
The presence of alloying elements which are relatively insoluble in zirconium under normal conditions, will generally result in "precipitates" forming within a zirconium matrix. Under equilibrium conditions, the matrix--which is a single phase--will contain the alloying elements at concentrations no higher than their solubility limit. The precipitates--which form a second phase--contain higher concentrations of the alloying elements. For example, the precipitates found in the Zircaloys are represented by chemical formulas such as Zr(Fe,Cr).sub.2 and Zr.sub.2 (Fe,Ni).
Cladding corrosion is a potential problem both in boiling water reactors (BWRs) and pressurized water reactors (PWRs). Corrosion in a BWR occurs in nodular or uniform formats on the zirconium cladding. Nodular corrosion is usually a porous, stoichiometric ZrO.sub.2 oxide forming on the surface of the cladding. It can rapidly cover the entire surface of pure zirconium, but it tends to form as small patches ("nodules" or "pustules") on the surface of the Zircaloys. Uniform corrosion is also a ZrO.sub.2 oxide forming on the surface of the cladding, but it usually contains a small excess of zirconium. As such, it contains excess electrons giving it a black or gray color and semiconductor properties.
Nodular or pustule corrosion is not inherently bad. However, where fuel in the reactor has longer life, nodular corrosion may concentrate. And where such concentrated nodular corrosion acts in conjunction with certain contaminants--such as copper ions--localized spalling and ultimately penetration of the cladding wall can occur.
Various approaches have been taken to minimize or eliminate nodular corrosion and the damage it can cause to cladding. In one widely used approach, the concentration of alloying elements (particularly iron, nickel, and chromium) in Zircaloy alloy is increased. This has been found to actually reduce the severity of nodular corrosion under reactor conditions. Unfortunately, increased concentrations of alloying elements also leads to increased rates of corrosion due to uniform corrosion. Even at such elevated rates, uniform corrosion has not been a significant problem in reactors operated under conditions common in the past. Today, however, it is increasingly common to operate reactors at high "burn-up" (i.e., to nearly complete consumption of the nuclear fuel). Under these conditions, the cladding is exposed to a neutron flux for long periods, a condition which tends to increase the degree of uniform corrosion. Thus, uniform corrosion can become a significant problem in modern reactor operation.
In another approach to nodular corrosion containment, the precipitates in the Zircaloy matrix are purposely made small (e.g., less than about 0.1 micrometer in diameter). They may be made small throughout the entire cross-section of the cladding or only in certain regions. For example, it is known to externally treat the outer water exposed surface of cladding with heating from a coil to produce a fine precipitate exterior surface. See Eddens et al. U.S. Pat. No. 4,576,654. Unfortunately, some research has suggested that small precipitates in the Zircaloy metal matrix can increase the danger of crack propagation in the cladding axial direction. See, for example, U.S. patent application No. 08/052,793, entitled ZIRCALOY TUBING HAVING HIGH RESISTANCE TO CRACK PROPAGATION and U.S. patent application Ser. No. 08/052,791, entitled METHOD OF FABRICATING ZIRCALOY TUBING HAVING HIGH RESISTANCE TO CRACK PROPAGATION, both of which were filed on Apr. 23, 1993, naming Adamson et al. as inventors, assigned to the assignee hereof, and are incorporated herein by reference for all purposes. These applications describe a cladding having a microstructure in which coarse precipitates predominate in the inner regions of the cladding and fine precipitates predominate in the outer regions of the cladding, the regions where corrosion is a problem.
Corrosion and cracking can both damage cladding, but they are fundamentally different phenomena. Cracking is a mechanical breaking or splitting of the cladding wall, while corrosion is an electrochemical conversion of the cladding metal into an oxide or other non-metallic compound. Cracks may be initiated by a variety of causes including mechanical stresses as well as corrosion. Once a crack is initiated, it may pose little problem, so long as it remains confined to a small area. However, if the crack propagates, the cladding can be breached and the fission material eventually contacts the coolant or moderator. Ultimately, this can lead to an expensive reactor outage.
The mechanical initiation of cracks can be attributed to various stresses in a conventional reactor. Cracks can start when debris such as wires or metallic shavings or particles find their way into reactor water that flows within the fuel bundles between the fuel rods. The debris may lodge at a fuel rod spacer adjacent the cladding wall. As a result, the debris vibrates or frets against the cladding wall under the influence of the passing steam/water mixture. Corrosion can be the source of initial crack propagation. Moreover, manufacturing defects can be the points of crack origin. Still further, crack propagation can start on the inside of the fuel rods in the corrosive high pressure environment present during in-service reactor life.
U.S. Pat. Nos. 4,200,492 and 4,372,817, to Armijo et al as well as Adamson U.S. Pat. No. 4,894,203--each of which is incorporated herein by reference for all purposes--suggest solutions to preventing crack initiation by including a barrier on the inside of the cladding. Cladding containing barriers are sometimes referred to as "composite" cladding or cladding having two distinct metallurgical layers.
Although it is highly desirable to prevent nodular corrosion of zirconium alloy cladding, it is also desirable to prevent uniform corrosion at high bum-up and to prevent axial crack propagation. There exists a need for cladding which is resistant to nodular corrosion while retaining resistance to uniform corrosion at high bum-up and axial crack propagation.